Nuclear reactors have their fuel contained in sealed cladding for the isolation of the nuclear fuel from the moderator/coolant system. The term cladding, as used herein, refers to a zirconium based alloy tube. The term precipitates, as used herein, refers to isolated structures distributed throughout the zirconium alloy matrix. These precipitates may or may not constitute intermetallics. Typically, these precipitates are uniformly distributed in the matrix--although they vary in size. Further, so-called fine precipitates (below 0.1 microns), can either be in the matrix format or the so-called two dimensional format where the precipitates occupy a sheet like layer near the outer surface of the zirconium alloy.
The cladding--typically in the order of 0.030 inches thick--is formed in the shape of a tube with the nuclear fuel contained typically in pellet form therein. These pellets are stacked in contact with one another for almost the entire length of each cladding tube, which cladding tube is in the order of 160 inches in length. Typically, the cladding tube is provided with springs for centering the fuel pellets and so-called "getters" for absorbing hydrogen and water vapor. Thereafter, the internal portions of the fuel rod are pressurized with various gases for optimum dissipation of gases produced from the fission reaction (and for increasing the thermal conductivity of the system), and sealed at both ends.
Zirconium and its alloys, under normal circumstances, are excellent nuclear fuel cladding since they have low neutron absorption cross sections and at temperatures below about 400.degree. C. are strong, ductile, extremely stable and nonreactive in the presence of demineralized water or steam. "Zircaloys" are a widely used family of corrosion-resistant zirconium alloy cladding materials. The Zircaloys are composed of 98-99% by weight zirconium, with the balance being tin, iron, chromium, and nickel. "Zircaloy-2" and "Zircaloy-4" are two widely-used zirconium-based alloys for cladding. (Zircaloy-4 omitting nickel).
Cladding corrosion is a potential problem both in boiling water reactors ("BWRs") and pressurized water reactors "PWRs"). For example, in a PWR, water does not boil--although in modern designs some boiling can occur at the top of some fuel rods. The oxygen level is relatively suppressed, being about 20 ppb. Hydrogen is injected and resident in the water moderator at about 200 ppb and utilized to suppress oxygen levels. Water pressure is in the range of 2000 psi with temperature ranging from 300.degree. C. to 380.degree. C. dependent upon the operating state of the reactor.
Corrosion in PWR cladding is uniform and related to precipitate size in the Zircaloy cladding. Small precipitates have been found to actually accelerate the uniform corrosion phenomena. Consequently, relatively large precipitate sizes are preferred in the PWR zirconium cladding.
In the radiation environment within the PWR, the precipitates dissolve and become smaller with exposure. To avoid accelerated uniform corrosion buildup, PWR cladding uniformly starts with large precipitate sizes--0.2 microns and above--to slow the formation of small size precipitates and the more rapid uniform corrosion that occurs with the small size precipitates.
In a BWR environment, water does boil. The oxygen level is relatively high, being about 200 ppb. Hydrogen may be injected for the stability of structural parts of the reactor, is effectively stripped off as a part of the boiling, and is resident in the water moderator in the range of 20 ppb. Water pressure is in the range of 1000 psi with temperature at 288.degree. C. being essentially a function of pressure and for the most part constant for all operating states of the reactor.
Corrosion in BWR cladding occurs in a nodular format. Uniform corrosion is also always present--but in the usual case not to a significant degree. Further, mineral and particle deposition occurs on the water exposed surface of the cladding. The combination of the corrosion and depositions can become fairly thick on the water exposed portions of the cladding.
Nodular corrosion is not inherently bad. However, where fuel in the reactor has longer life, nodular corrosion concentrates. Where such nodular corrosion becomes concentrated and acts in conjunction with other particles--such as copper ions--localized penetration of the cladding wall can occur.
Small precipitates have been found to actually suppress nodule formation. Consequently, it is desired to have small precipitates--below 0.1 microns--to inhibit formation of nodules. It is known in the prior art to externally treat the outer water exposed surface of cladding with heating from a coil to produce a fine precipitate exterior surface. See Eddens et al. U.S. Pat. No. 4,576,654.
In the radiation environment within the BWR, the precipitates dissolve and become smaller with radiation exposure. Nodular corrosion is inhibited by the small precipitates and by the alloying elements put in solution by the dissolution process.
Anneals of zirconium alloys have been used and can be summarized in terms of temperature ranges. Starting at low temperatures, anneals above 480.degree. C. effect stress relief, usually after working of the metal to achieve around 70% reduction in area. Anneals at about 576.degree. C. not only effect stress relief but also commence recrystallization of the metal. In such anneals, maximum ductility of the material is achieved. Finally, anneals substantially above 576.degree. C. effect crystal growth--generally softening the metal.
In the prior art, the heat treatment for PWR cladding has included high temperature anneals with slow quenches (less than 50.degree./second) to preserve large precipitate sizes. Conversely, the heat treatment for BWR cladding has included low temperature anneals with fast quenches (greater than 50.degree./second) to produce small precipitate sizes.
Corrosion and cracking can both damage cladding, but they are fundamentally different phenomena. Cracking is a mechanical breaking or splitting of the cladding wall, while corrosion is an electrochemical conversion of the cladding metal into an oxide or other non-metallic compound. Cracks may be initiated by a variety of causes including mechanical stresses. Once a crack is initiated, it may pose little problem, so long as it remains confined to a small area. However, if the crack propagates, the enlarged opening can permit the fission material to eventually contact the coolant or moderator and result in significant release of radioactive species to the coolant. Ultimately, this can lead to an expensive reactor outage.
Regarding cracking in the interior of the sealed cladding tube, brittle splitting of such cladding may occur due to the combined interactions between the nuclear fuel, the cladding, and the fission products produced during the nuclear reaction. It has been found that this undesirable performance is due to localized mechanical stresses on the fuel cladding resulting from differential expansion and friction between the fuel and the cladding. These localized stresses and strain in the presence of specific fission products, such as iodine and cadmium, are capable of producing cladding failures by phenomena known as stress corrosion cracking and liquid metal embrittlement. Other phenomena such as local hydriding of the cladding and the presence of oxygen, nitrogen, carbon monoxide, and carbon dioxide can assist cladding failure and lead to rod cracking.
U.S. Pat. Nos. 4,200,492 and 4,372,817 to Armijo et al as well as Adamson U.S. Pat. No. 4,894,203 suggest solutions to preventing crack initiation by including a barrier on the inside of the cladding. Such cladding containing an introduced barrier are sometimes referred to as "composite" cladding or cladding having two distinct metallurgical layers.
Although it is highly desirable to prevent crack initiation, in the event a crack forms, its propagation is to be avoided.
There exists a need, especially for a BWR environment, for cladding which is resistant to axial crack propagation. There also, exists a need for cladding which, in combination, is resistant to axial crack propagation, crack initiation and corrosion.